November 12, 2004 Mr. Bryce L. Shriver President, PPL Generation, LLC and

November 12, 2004
Mr. Bryce L. Shriver
President, PPL Generation, LLC and
Chief Nuclear Officer
PPL Generation, LLC
2 North Ninth Street
Allentown, PA 18101
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION - NRC INTEGRATED
INSPECTION REPORT 05000387/2004004 AND 05000388/2004004
Dear Mr. Shriver:
On September 30, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Susquehanna Steam Electric Station, Units 1 and 2. The enclosed
integrated inspection report presents the results of that inspection, which was discussed with
Mr. B. McKinney, Vice President - Nuclear Site Operations and other members of your staff on
October 7, 2004.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commission’s rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents three findings of very low safety significance (Green). One of the
findings was determined to involve a violation of NRC requirements. However, because of the
very low safety significance and because the issue was entered into your corrective action
program, the NRC is treating this finding as non-cited violation (NCV), consistent with Section
VI.A of the NRC Enforcement Policy. Additionally, one licensee-identified violation, which was
determined to be of very low safety significance, is listed in this report. If you contest the NCV
in this report, you should provide a response within 30 days of the date of this inspection report,
with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator Region
I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at the Susquehanna Steam Electric Station.
In accordance with 10 CFR 2.390 of the NRC’s "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publically Available Records (PARS) component of the NRC’s document
Mr. Bryce L. Shriver
2
system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
If you have any questions please contact me at (610) 337-5209.
Sincerely,
/RA/
Mohamed M. Shanbaky, Chief
Projects Branch 4
Division of Reactor Projects
Docket Nos. 50-387; 50-388, 72-28
License Nos. NPF-14, NPF-22
Enclosure:
Inspection Report 05000387/2004004 and 05000388/2004004
Attachment: Supplemental Information
cc w/encl:
J. H. Miller, Executive Vice-President and COO - PPL Services
B. T. McKinney, Vice President - Nuclear Site Operations
R. A. Saccone, Vice President - Nuclear Operations for PPL Susquehanna LLC
A. J. Wrape, III, General Manager - Performance Improvement and Oversight
T. L. Harpster, General Manager - Plant Support
K. Roush, Manager - Nuclear Training
G. F. Ruppert, General Manager - Nuclear Engineering
J. M. Helsel, Manager - Nuclear Operations
R. D. Pagodin, Manager - Station Engineering
J. E. Krais, Manager - Nuclear Design Engineering
T. Mueller, Manager - Nuclear Maintenance
D. Glassic, Manager - Work Management
R. E. Smith, Jr., Radiation Protection Manager
N. Grisewood, Manager - Corrective Action & Assessments
D. F. Roth, Manager - Quality Assurance
R. R. Sgarro, Manager - Nuclear Regulatory Affairs
R. Ferentz, Manager - Nuclear Security
W. E. Morrissey, Supervisor - Nuclear Regulatory Affairs
M. H. Crowthers, Supervising Engineer
L. A. Ramos, Special Office of the President
B. A. Snapp, Esquire, Associate General Counsel, PPL Services Corporation
R. W. Osborne, Allegheny Electric Cooperative, Inc.
Board of Supervisors, Salem Township
J. Johnsrud, National Energy Committee
Supervisor - Document Control Services
D. Allard, Director, Pennsylvania Bureau of Radiation Protection
Commonwealth of Pennsylvania (c/o R. Janati, Chief, Division of Nuclear Safety,
Pennsylvania Bureau of Radiation Protection)
Mr. Bryce L. Shriver
3
Distribution w/encl: via e-mail
S. Collins, RA
J. Wiggins, DRA
M. Shanbaky, DRP
A. Blamey, DRP - SRI Susquehanna
J. Richmond, DRP - RI Susquehanna
F. Jaxheimer, DRP - RI Susquehanna
S. Farrell, DRP - Susquehanna OA
J. Jolicoeur, RI OEDO
R. Laufer, NRR
R. Guzman, NRR
R. Clark, PM, NRR (Backup)
Region I Docket Room (with concurrences)
DOCUMENT NAME: E:\Filenet\ML043170533.wpd
After declaring this document “An Official Agency Record” it will be released to the Public.
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with
attachment/enclosure "N" = No copy
OFFICE RI/DRP
RI/DRP
NAME
DATE
MShanbaky/MMS
11/10/04
ABlamey/AJB
11/10/04
OFFICIAL RECORD COPY
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos.
50-387, 50-388, 72-28
License Nos.
NPF-14, NPF-22
Report No.
05000387/2004004, 05000388/2004004
Licensee:
PPL Susquehanna, LLC
Facility:
Susquehanna Steam Electric Station
Location:
769 Salem Boulevard
Berwick, PA 18603
Dates:
July 1, 2004 through September 30, 2004
Inspectors:
A. Blamey, Senior Resident Inspector
J. Richmond, Resident Inspector
F. Jaxheimer, Resident Inspector
N. McNamara, EP Inspector
J. Furia, Sr. Health Physicist
T. Burns, Reactor Inspector
J. Wray, Health Physicist
Approved by:
Mohamed M. Shanbaky, Chief
Projects Branch 4
Division of Reactor Projects
i
Enclosure
CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R01 Adverse Weather . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R04 Equipment Alignments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R06 Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R11 Licensed Operator Requalification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R12 Maintenance Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R13 Maintenance Risk Assessments & Emergent Work Evaluation . . . . . . . . . . . . 10
1R14 Personnel Performance During Non-Routine Plant Evolutions . . . . . . . . . . . . 10
1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
1R16 Operator Work-Around Cumulative Review . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1R17 Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
1R19 Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1R23 Temporary Plant Modification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1EP4 Emergency Action Level (EAL) and Emergency Plan Changes . . . . . . . . . . . 17
1EP6 Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 18
2OS2 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
2OS3 Radiation Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
4OA3 Event Follow-up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4OA4 Cross Cutting Aspects of Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
4OA5 Other . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
4OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
4OA7 Licensee-identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
ATTACHMENT: SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
KEY POINT OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . .
LIST OF BASELINE INSPECTIONS PERFORMED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
ii
A-1
A-1
A-1
A-2
A-2
A-4
Enclosure
SUMMARY OF FINDINGS
IR 05000387/2004004, 05000388/2004004; 07/01/2004 - 09/30/2004; Susquehanna Steam
Electric Station, Units 1 and 2; Maintenance Implementation, Post Maintenance Test,
Identification and Resolution of Problems.
The report covered a 3-month period of inspection by resident inspectors and announced
inspections by a regional emergency preparedness inspector, senior health physicist, health
physicist, and a reactor inspector. One non-cited violation (NCV) of very low safety
significance, two Green findings and one unresolved item were identified. The significance of
most findings are indicated by their color (Green, White, Yellow, Red) using Inspection Manual
Chapter 0609 "Significance Determination Process" (SDP). Findings for which the SDP does
not apply may be Green or be assigned a severity level after NRC management review. The
NRC’s program for overseeing the safe operation of commercial nuclear power reactors is
described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.
A.
NRC-Identified Findings
Cornerstone: Initiating Events
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Green. The inspectors identified a finding because PPL did not complete an
evaluation of the condition of Unit 2 transformer 2X270 as required by a station
procedure before energizing the transformer. Not completing the evaluation
allowed a degraded transformer to be returned to service. The transformer
faulted shortly after being placed in service which resulted in a loss of main
condenser vacuum.
This finding is greater than minor because it adversely impacts the equipment
performance attribute of the Initiating Events cornerstone and the finding
adversely affected the cornerstone objective, in that, it is associated with an
event that upset plant stability. This finding was considered to have very low
safety significance (Green), using phase 1 of the significance determination
process. The failure of transformer 2X270 did not increase the likelihood of an
LOCA initiator, and did not increase the likelihood of a reactor trip and the
likelihood that mitigation functions would be lost. In addition, the finding did not
increase the likelihood of a fire or flood event.
A contributing cause of this finding is related to the Human Performance crosscutting area because PPL did not complete the required retest and engineering
evaluation of transformer 2X270 prior to energizing the transformer.
(Section 1R19)
Cornerstone: Mitigating Systems
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Green. The inspectors identified a non-cited violation of 10CFR 50.65 paragraph
(b)(2) of the Maintenance Rule, because PPL did not scope the Unit 1 and Unit 2
reactor building (RB) equipment and floor drain systems (EFDS) into the
Maintenance Rule program and as a result did not demonstrate the effectiveness
iii
Enclosure
of preventive maintenance for the RB EFDS. The inclusion of the RB EFDS in
the scope of the monitoring program was necessary because the RB EFDS are
relied upon to mitigate internal flooding events. Failure of the EFDS to function
could have prevented safety-related structures, systems and components from
fulfilling their safety-related function.
This finding was more than minor because it had greater significance than
similar issues described in the NRC Inspection Manual Chapter 0612, “Examples
of Minor Issues,” Section 1.h and 1.i. In addition, the RB EFDS’s performance is
associated with the Equipment Performance attribute of the Mitigating Systems
cornerstone and adversely affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events
to prevent undesirable consequences. On August 18, 2004, the Unit 1 RB EFDS
was unable to pass 80 gpm as assumed in the Final Safety Analysis Report
during an overflow of the reactor water cleanup backwash receiving tank.
Inspectors identified that system performance problems were such that a
Maintenance Rule (a)(2) demonstration could not be justified. This finding was
considered to have very low safety significance because the finding did not
contribute to an actual loss of mitigation equipment functions, and did not
increase the likelihood of a fire or flooding event.
A contributing cause of this finding was related to Problem Identification and
Resolution cross-cutting area. PPL had eleven previous EFDS blockages and
the evaluation of those events did not recognize that portions of the non-safety
related EFDS were relied upon to mitigate accidents or transients. Therefore,
PPL did not monitor the EFDS under the maintenance rule and this contributed
to the degradation of the RB EFDS. (Section 1R12).
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Green. A finding of low safety significance was identified because PPL did not
adequately evaluate and correct a degraded condition associated with the high
engine operating temperatures and repetitive overheating of the diesel driven fire
pump (DFP) which occurred following engine shutdown.
This issue is greater than minor because it affected the Mitigating Systems
cornerstone objective of ensuring the availability, reliability and capability of
systems that respond to initiating events to prevent undesirable consequences.
This finding is of very low safety significance, based on a Phase 1 significance
determination process evaluation, because the finding did not result in the loss of
a function of equipment designed as risk significant for greater than 24 hours
and the finding does not increase the potential or risk of a seismic event , flood
or severe weather event.
A contributing cause of this finding is related to the Problem Identification and
Resolution (PI&R) cross-cutting area. PPL did not sufficiently evaluate the
condition to identify and correct the reduced cooling water flow to the DFP
engine. This resulted in ineffective corrective actions because the DFP was
removed from service several times without taking action to correct the DFP high
engine coolant temperature issue. (Section 4OA2.3)
iv
Enclosure
B.
Licensee Identified Violation
A violation of very low safety significance, which was identified by PPL, has been
reviewed by the inspectors. Corrective actions taken or planned by PPL have been
entered into PPL’s corrective action program. This violation and corrective actions are
listed in Section 4OA7 of this report.
v
Enclosure
REPORT DETAILS
Summary of Plant Status
Susquehanna Steam Electric Station (SSES) Unit 1 began the inspection period at full power.
On August 22, 2004, reactor power was reduced to 70 percent power due to a problem with the
“B” reactor feedwater pump turbine control valve operator linkage. This problem with the
linkage was corrected and Unit 1 was returned to 100 percent power on August 25, 2004. On
September 10, 2004, reactor power was reduced to 70 percent power for planned maintenance
on feedwater heaters. Maintenance activities were completed and Unit 1 was returned to 100
percent power on September 13, 2004. Unit 1 was operated at or near full power for the
remainder of the inspection period.
Unit 2 was operating at or near full power at the beginning of the inspection period. On
July 29, 2004, reactor power was rapidly reduced to 78 percent power due to the loss of a 13.8
KV stepdown transformer 2X270. The power loss impacted condenser air removal capability
reducing condenser vacuum. Alternate equipment was put in service and Unit 2 returned to
100 percent power on July 30, 2004. Unit 2 reactor power was reduced to 70 percent on
September 18, 2004 for a control rod sequence exchange and returned to full power on
September 19, 2004. Unit 2 was operated at or near full power for the remainder of the
inspection period, with exceptions for brief power reductions to support control rod pattern
adjustments or to support transmission and distribution limitations (minimum generation alerts).
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather (71111.01- 3 Samples)
a.
Inspection Scope
The inspectors reviewed PPL’s preparations for adverse weather conditions and
performed plant walkdowns for selected structures, systems, and components. The
walkdowns and reviews were conducted to determine the adequacy of PPL's weather
protection activities and system features. The inspectors reviewed and evaluated plant
conditions related to hot weather and storm preparations. The inspectors reviewed the
procedures for hot weather and high wind protection of the associated systems. This
inspection activity represented three samples. The areas, components, and documents
reviewed included:
Structures, Systems, and Components
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Onsite / Offsite electrical power system, hurricane Charley, August 13, 2004
Onsite / Offsite electrical power system, hurricane Ivan, September 15, 2004
Closed cooling water systems, August 13, 2004, hot weather preparation
Enclosure
2
Procedures and Documents
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b.
NDAP-00-0024, “Winter Operation Preparations and Severe Weather Operation”
ON-000-002, “Natural Phenomena”
ON-000-005, “Hot Weather”
CR 549002, Unit 2 ISO Phase Flux Plates Continue to be Hot (422EF)
Findings
No findings of significance were identified.
1R04 Equipment Alignments (71111.04Q - 6 samples)
1.
a.
Partial System Walkdowns
Inspection Scope
The inspectors performed partial system walkdowns to verify system and component
alignment and to note any discrepancies that would impact system operability. The
inspectors verified selected portions of redundant or backup systems or trains were
available while certain system components were out of service. The inspectors
reviewed selected valve positions, electrical power availability, and the general condition
of major system components. This inspection activity represented six samples. The
walkdowns included the following systems:
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b.
Fire water system with alternate suction alignment and diesel fire pump
troubleshooting in progress
Unit 2 division II RHR with division I RHR in SOW, August 11, 2004
Unit 1 division I core spray (following the RWCU overflow event - section 1R06)
Unit 2 turbine building closed cooling water (TBCCW) during and post
maintenance on service water pressure safety valve (PSV) to B cooler,
September 7, 2004
Unit 1, 4 KV emergency bus alignment during 4 KV emergency bus 1C feeder
breaker replacement, September 7, 2004
ESW following flow balance testing, September 16, 2004
Findings
No findings of significance were identified.
Enclosure
3
1R05 Fire Protection (71111.05Q - 11 Samples)
1.
a.
Routine Plant Area Observations
Inspection Scope
The inspectors reviewed PPL's fire protection program to determine the required fire
protection design features, fire area boundaries, and combustible loading requirements
for selected areas. The inspectors walked down those areas to assess PPL’s control of
transient combustible material and ignition sources, fire detection and suppression
capabilities, fire barriers, and any related compensatory measures to assess PPL's fire
protection program in those areas. The inspectors reviewed the respective pre-fire
action plan procedures for the inspected areas. This inspection activity represented
eleven samples. The inspected areas included:
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b.
Unit 2 reactor core isolation cooling (RCIC) room, FP-213-239
Unit 2 high pressure coolant injection (HPCI) room, FP-213-238
Reactor building 719' elevation, 3-hour fire barriers between Units 1 and 2, CR
582588
Unit 1 Reactor Building 670' elevation, Fire Zones 1-2B and 1-2D
Unit 1 Reactor Building 683' elevation, Fire Zone 1-3A
Control Structure Fire Zones 0-26M and 0-26R, CO2 Manual Spurt Suppression
Unit 2 Reactor Building 683' elevation, Fire Zones 2-3B-W and 2-3C-W
Control Room, Tech Support Center and adjacent areas, FP-013-156
Unit 1 battery rooms and DC distribution panel areas, Fire Zones 0-28j, k, l & m.
Unit 1 & 2 TBCCW pump and heat exchanger area, Fire Zone 0-21A
Unit 1 Switchgear rooms, 719' elevation, FP-113-115
Findings
No findings of significance were identified.
1R06 Flood Protection Measures (71111.06 - 2 Samples)
1.
a.
Internal Flood Protection
Inspection Scope
On August 18, 2004, during an extended backwash evolution on the reactor water
cleanup demineralizer at Susquehanna Unit 1, the backwash receiving tank overflowed
and placed approximately 1500 gallons of contaminated water into the reactor building
equipment floor drain system. The drain header became blocked due to the resin from
the receiving tank and rust that was displaced from the inside of the drain piping. The
water flowed up and out of the blocked drains on a lower elevation, across the floor and
down into the division II core spray and the high pressure coolant injection system
compartments. The water entered these compartments by flowing though gaps
between the equipment hatch floor plugs and the floor that were not sealed. The
Enclosure
4
division II core spray and high pressure injection system compartments both had
approximately 2 inches of water on the floor.
The inspectors reviewed the event and its impact on safety related equipment. The
inspectors compared the plant response to the expected plant response based on the
flood protection design features specified in the Final Safety Analysis Report (FSAR),
engineering analysis. The inspectors conducted Unit 1 walkdowns to independently
assess the leakage paths, water accumulation, and the operability of equipment that
was in the affected areas of the spill. This inspection activity represented one sample.
Documents reviewed during the inspection are listed in the attachment, “List of
Documents Reviewed.”
b.
Findings
Large equipment hatches are installed in the ceilings of all the emergency core cooling
systems (ECCS) rooms to allow for removal of equipment during maintenance activities.
The equipment hatches are normally closed by the use of large equipment hatch plugs
which are placed in these openings. PPL reviewed the design of the equipment hatch
plugs after the August 18, 2004, event and determined that the plugs would not provide
a water tight seal between the plug and the surrounding floor, even though this was
specified in the FSAR, Section 3.4. In addition, the plant specific analysis did not
consider any leakage through the equipment hatch plugs. The analysis accounted for
leakage around doors and through the floor drain system but not through the equipment
hatch plugs. Therefore, the analysis implied that the equipment hatch plugs were water
tight. PPL is currently reanalyzing the flooding issue to determine if the equipment
hatch floor plugs must be water tight. This analysis is being performed under condition
report 600070. This issue is unresolved pending PPL’s completion of the flooding
analysis and the NRC review of this analysis. This finding is identified as URI
05000387/2004004-01, “Equipment Hatch Floor Plugs are not Watertight as
Indicated in the FSAR.”
2.
a.
External Flood Protection Measures
Inspection Scope
The inspectors reviewed PPL's external flood analysis, flood mitigation procedures, and
design features to verify whether they were consistent with the PPL design
requirements. The inspectors walked down selected risk significant plant areas,
including the moats and surrounding areas for large on-site tanks. The inspectors
evaluated the condition and adequacy of flood detectors, sump pumps, sump level
alarm circuits, and other flood protection design features to assess whether the flood
protection design features were adequate and operable. During the walk downs, the
inspectors also evaluated whether there were any unidentified or unanalyzed sources of
flooding, including holes and unsealed penetrations in floors and walls. This inspection
activity represented one sample. The specific areas included:
Enclosure
5
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Turbine and reactor building protection from circulating water flume rupture
“A” through “D” emergency diesel generator rooms
“E” emergency diesel generator building
The inspectors reviewed PPL's flood mitigation procedures, flood alarm response
procedures, and selected preventive maintenance tasks for flood detectors and flood
barriers to evaluate whether component functionality was routinely verified. In addition,
the inspectors reviewed PPL's corrective action program, including system health
reports, and interviewed selected maintenance personnel to verify whether previous
flood related issues had been appropriately identified, evaluated, and resolved. The
following procedures were included in the review:
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b.
NE-94-001, Section 5.2, “Susquehanna IPE for External Events - Floods”
FSAR Section 2.4.2, “Hydrologic Engineering - Floods”
FSAR Section 3.4, “Water Level (Flood) Design”
EC-RISK01024, “External Flood Effects”
CR# 598927, Turbine Building Overhead Door Opening
Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07 - 2 Samples)
a.
Inspection Scope
The inspectors reviewed PPL’s inspection, cleaning, and maintenance activities, and
reviewed PPL’s evaluation of the as-found conditions for the Unit 2 “A” residual heat
removal (RHR) room cooler (2E230A) and the Unit 2 “A” high pressure coolant injection
(HPCI) room cooler (2E229A). The inspectors verified whether PPL properly evaluated
the results to identify adverse trends and ensure adequate heat transfer capabilities.
The inspectors compared their observations against PPL’s procedures and
specifications to assess whether the heat exchangers were capable of performing their
safety function under design basis accident conditions. This inspection activity
represented two samples. The inspectors' review included the following documents:
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b.
WO 541964, Unit 2 Division I RHR Room Cooler clean and inspect
WO 541922, Unit 1 Division I HPCI Room Cooler clean and inspect
EC-CHEM-1018, Justification for the Assurance of Adequate Heat Removal
Capabilities Using the Susquehanna Heat Exchanger Preventive Maintenance
Program
Findings
No findings of significance were identified.
Enclosure
6
1R11 Licensed Operator Requalification (71111.11Q - 1 Sample)
1.
a.
Routine Licensed Operator Requalification
Inspection Scope
On August 25, 2004, the inspectors observed a simulator sessions of operator
requalification training. The inspectors compared the actions taken during the simulator
scenario to classroom objectives, and compliance with Technical Specifications, NRC
orders, and emergency operating procedures. The inspectors' evaluation focused on
the operating crew’s satisfactory completion of crew critical tasks, and satisfactory
implementation of the emergency plan and emergency action level (EAL) classifications
for the simulated plant conditions. Critical tasks are operational limits placed on key
reactor plant and containment parameters that will ensure safety margins are
maintained during the simulated malfunctions. The review included a comparison of the
simulator’s ability to model the actual plant performance. The inspectors also evaluated
PPL’s critique of the operators' performance to identify deficiencies in operator training.
This inspection activity represented one sample. The observed training scenario
included Security Events, Station Procedure ON-000-010.
b.
Findings
No findings of significance were identified.
1R12 Maintenance Implementation (71111.12Q - 5 Samples)
1.
a.
Routine Review of Maintenance Implementation
Inspection Scope
The inspectors evaluated PPL’s work practices and follow-up corrective actions for
selected system, structure, or component (SSC) issues to assess the effectiveness of
PPL's maintenance activities. The inspectors reviewed the performance history of those
SSCs and assessed PPL’s extent of condition determinations for these issues with
potential common cause or generic implications to evaluate the adequacy of PPL’s
corrective actions. The inspectors reviewed PPL's problem identification and resolution
actions for these issues to evaluate whether PPL had appropriately monitored,
evaluated, and dispositioned the issues in accordance with PPL procedures and the
requirements of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of
Maintenance." In addition, the inspectors reviewed selected SSC classification,
performance criteria and goals, and PPL's corrective actions that were taken or planned,
to verify whether the actions were reasonable and appropriate. This inspection activity
represented four samples. The following issues were reviewed:
Equipment Issues
C
Loss of Remote Position Indication for SV-22651
Enclosure
7
C
C
C
C
Loss of Remote Position Indication for HV-15725
Loss of Remote Position Indication for HV-22603
Emergency lighting system 07 return to A2 status and functional failure of
lighting battery 20135
Reactor Building Floor and Equipment Drains
Procedures and Documents
C
C
C
C
C
b.
NDAP-QA-0413, “Maintenance Rule Program”
Unexpected Loss of DC Power to the Inboard MSIV and SV-22651, (CR 597331)
Loss of Open Indication for Containment Isolation Valve HV-22603, (CR 597822)
Primary Containment Instrument Gas Maintenance Rule Basis Document
Emergency Lighting Maintenance Rule Basis Document
Findings
No findings of significance were identified.
2.
a.
Equipment Floor Drain System
Inspection Scope.
The inspectors evaluated PPL’s work practices and preventive maintenance activities for
the reactor building (RB) equipment and floor drain systems (EFDS) to assess the
effectiveness of PPL’s maintenance activities. The inspectors reviewed the
performance history of the RB EFDS to assess PPL’s problem identification and to
evaluate whether PPL had appropriately monitored, evaluated and dispositioned issues
in accordance with PPL procedures and the requirements of 10 CFR 50.65,
“Requirements for Monitoring the Effectiveness of Maintenance.” The inspectors
reviewed the associated system design basis, including the Final Safety Analysis Report
(FSAR) and the RB internal flood design calculations, to assess the adequacy of PPL’s
actions. In addition, the inspectors performed field walkdowns and interviewed PPL
staff to verify whether the identified actions were appropriate and to verify that known
performance problems were included in the corrective action process. This inspection
activity represented one sample.
b.
Findings
Introduction. A Green non-cited violation (NCV) of 10 CFR 50.65 (b)(2) was identified
because PPL did not scope an accident and transient mitigation function of the EFDS
into the Maintenance Rule monitoring program. The inspectors concluded that not
having the EFDS function scoped into the monitoring program allowed the deterioration
of system performance such that the Unit 1 RB EFDS could not perform the intended
design function on August 18, 2004 (Section 1R14).
Description. The NRC identified that PPL did not correctly scope the RB EFDS into the
maintenance rule program. The Maintenance Rule, 10 CFR 50.65 paragraph (b)(2),
Enclosure
8
requires that systems whose failure to function as designed could prevent safetyrelated structures, systems and components from fulfilling their safety-related functions
should be scoped into the maintenance rule program. Since the RB EFDS was not
scoped into the maintenance monitoring program, PPL did not establish performance
criteria to demonstrate the effectiveness of the RB EFDS maintenance. Therefore, PPL
did not identify less than adequate preventive maintenance and the system failed to
meet design requirements on August 18, 2004, when the reactor water cleanup
(RWCU) backwash receiving tank overflowed and the EFDS became blocked.
The RB EFDS are required to mitigate internal flooding events such as moderate and
high energy line breaks; as well as, fire deluge system actuations to prevent this water
from impacting safety-related equipment. The Susquehanna FSAR, Section 3.4, “Water
Level (Flood) Design,” states the capacity of a single floor drain is approximately 80
gpm. The PPL design calculation EC-FLOD-0500, assumed 80 gpm per floor drain and
a maximum of 200 gpm for each reactor building drain header. During the August 18,
2004, event one RB EFDS header became blocked which prevented the required flow
assumed in the FSAR and supporting analysis.
The inspectors also identified that PPL had previously identified occurrences of EFDS
blockage. During the past three years there were eleven occurrences of backed up
EFDS in the Unit 1 and Unit 2 reactor buildings. The evaluation of these occurrences
did not recognize that portions of the non-safety related floor drain system are relied
upon to mitigate accidents or transients and did not specify adequate corrective actions
to prevent recurrence of the EFDS blockage. PPL had established and performed a
routine task to unplug portions of the reactor building and turbine building EFDS.
However, the preventive maintenance tasks were scheduled to ensure a functional drain
system before the beginning of a refueling outage, not as a deliberate measure to
maintain the systems functional during the plant operating cycle. Therefore, the
inspectors determined that the maintenance history and associated system performance
problems indicated that PPL had missed opportunities to place these components in the
maintenance rule scope and as a result did not effectively maintain the function of the
EFDS through appropriate preventive maintenance.
Analysis. The finding was a performance deficiency because PPL did not scope
sections of the system into the Maintenance Rule Program and consequently did not
provide an adequate Maintenance Rule (a)(2) demonstration for the RB EFDS.
Traditional enforcement does not apply because the issue did not have any actual safety
consequence or potential for impacting the NRC’s regulatory function and was not the
result of any willful violation of NRC requirements. The finding was more than minor
because it had greater significance than similar issues described in the NRC Manual
Chapter 0612, “Examples of Minor Issues”, Section 1.h and 1.i. In addition, the RB
EFDS’s performance is associated with the Equipment Performance attribute of the
Mitigating Systems cornerstone and adversely affected the cornerstone objectivity to
ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences because on August 18, 2004, the Unit 1
RB EFDS was unable to pass 80 gpm as assumed in the Final Safety Analysis Report
during an overflow of the reactor water cleanup backwash receiving tank.
Enclosure
9
This finding was considered to have very low safety significance (Green) using the
NRC’s Significance Determination Process (SDP) for Reactor Inspection Findings for AtPower situations because the finding did not contribute to a loss of mitigation equipment
functions and did not increase the likelihood of a fire or flooding event.
A contributing cause of this finding was related to Problem Identification and Resolution
cross-cutting area. PPL had eleven previous EFDS blockages and the evaluation of
those events did not recognize that portions of the non-safety related EFDS were relied
upon to mitigate accidents or transients. Therefore, PPL did not monitor the EFDS
under the maintenance rule and this contributed to the degradation of the RB EFDS.
Enforcement. 10CFR 50.65 (b)(2) requires, in part, that the scope of the monitoring
program specified in paragraph (a)(1) include non-safety related structures, systems
and components whose failure can prevent safety-related structures, systems and
components from fulfilling their safety-related functions.
10 CFR (a)(2) states that “Monitoring as specified in paragraph (a)(1) of this section is
not required where it has been demonstrated that the performance or condition of a
structure, system, or component is being effectively controlled through the performance
of appropriate preventive maintenance, such that the structure, system, or component
remains capable of performing its intended function.”
Contrary to the above, PPL did not include sections of the RB EFDS in the scope of the
monitoring program as specified in 10 CFR 50.65(b)(2). The inclusion of the RB EFDS
in the scope of the monitoring program was necessary because the system is utilized in
the mitigation of internal flooding events. As a result of not scoping, a RB EFDS
function into the monitoring program, PPL did not effectively control the performance or
condition of the system through appropriate preventive maintenance, as required by
10CFR 50.65(a)(2). The RB Equipment and Floor drain system did not perform its
intended function on August 18, 2004. System performance demonstrated that the
system would not have functioned for more severe internal flooding events to protect
safety-related equipment. Therefore, the inspectors concluded that as a result of this
finding the RB EFDS system had not been effectively controlled through the
performance of appropriate preventive maintenance and, as a result, a Maintenance
Rule (a)(2) demonstration could not be justified.
Because this finding was of very low safety significance and it was entered into the PPL
corrective action program, this finding is being treated as non-cited (NCV), consistent
with Section VI.A of the NRC Enforcement Policy. NCV 05000387/2004004-02,
“Reactor Building Floor and Equipment Drains Not Fully Scoped into the
Maintenance Rule.”
Enclosure
10
1R13 Maintenance Risk Assessments & Emergent Work Evaluation (71111.13 - 5 Samples)
a.
Inspection Scope
The inspectors reviewed the assessment and management of selected maintenance
activities to evaluate the effectiveness of PPL's risk management for planned and
emergent work. The inspectors compared the risk assessments and risk management
actions to the requirements of 10 CFR 50.65(a)(4) and the recommendations of
NUMARC 93-01 Section 11, "Assessment of Risk Resulting from Performance of
Maintenance Activities." The inspectors evaluated the selected activities to determine
whether risk assessments were performed when required and appropriate risk
management actions were identified.
The inspectors reviewed scheduled and emergent work activities with licensed operators
and work-coordination personnel to verify whether risk management action threshold
levels were correctly identified. In addition, the inspectors compared the assessed risk
configuration to the actual plant conditions and any in-progress evolutions or external
events to evaluate whether the assessment was accurate, complete, and appropriate for
the issue. The inspectors performed control room and field walkdowns to verify whether
the compensatory measures identified by the risk assessments were appropriately
performed. This inspection activity represented five samples. The selected
maintenance activities included:
C
C
C
C
C
b.
Troubleshooting triple notch of control rod 18-23, TP-055-010
Removal of division I ESW, draining and refill and venting, AR 592531
RCIC remote shutdown panel switch replacement, RLWO 595679
Loss of remote position indication for HV-22603, CR 597822
1A20301 breaker replacement (alternate power to 1C ES BUS), CR 603513
Findings
No findings of significance were identified.
1R14 Personnel Performance During Non-Routine Plant Evolutions (71111.14 - 2 Samples)
3.
a.
Unit 1 Reactor Water Cleanup Backwash Receiving Tank Overfill
Inspection Scope
The inspectors reviewed the station response to this August 18, 2004 overflow of
approximately 1500 gallons of contaminated water from an extended backwash
evolution on the Unit 1 reactor water cleanup demineralizer on August 18, 2004. This
event resulted in unexpectedly placing contaminated water into the division II core spray
and the HPCI system compartments. The inspectors reviewed the initial operator and
radiation protection response to this event. The inspectors conducted Unit 1 walkdowns
to independently assess the leakage paths, water accumulation, and additional water
sources. This event is discussed further in Section 1R06, “Flood Protection Measures,”
Enclosure
11
and Section 1R12, “Maintenance Implementation,” of this report. This inspection activity
represented one sample.
b.
Findings
No findings of significance were identified.
2.
a.
Unit 2 13.8 KV Stepdown Transformer (2X270) Electrical Fault
Inspection Scope
The inspectors reviewed the July 29, 2004 plant transient and operator response
following the electrical fault on transformer 2X270. This fault resulted in isolation of the
condenser offgas system, which reduced condenser vacuum and require the operators
to reduce reactor power to 78 percent. Specifically, the inspectors reviewed the
operator actions to stabilize the plant and the ascension to full reactor power. The
inspectors compared the equipment and operator responses for the July 29 transient to
the similar event that occurred on April 28, 2004. The inspectors reviewed and
evaluated PPL’s April 28, 2004, event root cause analysis and the corrective actions that
were taken to prevent recurrence. This event is discussed further in section 1R19, “Post
Maintenance Testing,” of this report. This inspection activity represented one sample.
b.
Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15 - 6 Samples)
a.
Inspection Scope
The inspectors reviewed operability determinations that were selected based on risk
insights, to assess the adequacy of the evaluations, the use and control of
compensatory measures, and compliance with the Technical Specifications. In addition,
the inspectors reviewed the selected operability determinations to verify whether the
determinations were performed in accordance with NDAP-QA-0703, "Operability
Assessments." The inspectors used the Technical Specifications, Technical
Requirements Manual, Final Safety Analysis Report (FSAR), and associated Design
Basis Documents as references during these reviews. This inspection activity
represented six samples. The issues reviewed included:
C
C
C
C
C
C
Replacement of leaking ESW supply and return valves to “E” EDG, CR 544629,
592531
Unit 2 control rod double notch and lockup, CR 591003, 591010, 591011
CIG Header pressure below SO-100-006 required valve of 90 psig, CR 597117
No Flow indicated during SLC pump surveillance SO-253-004, CR 599727
Reactor building internal flood/physical separation (Floor Plugs), CR 600070
Breaker 1A20309 failed to auto open on closure of 1A20301, CR 603513
Enclosure
12
b.
Findings
No findings of significance were identified.
1R16 Operator Work-Around Cumulative Review (71111.16 - 3 Samples)
a.
Inspection Scope
The inspectors reviewed the loss of the safety parameter display system (SPDS) and
the loss of the system particulate, iodine, and nobel gas (SPING) terminals in the control
room and the technical support center to determine how the effected system would
impact the emergency response staff’s assessment capability during an event. In
addition, the inspectors reviewed significant control room deficiencies, status control
tags, and selected corrective action reports to determine whether the functional
capability of a system or staff emergency response actions would be affected. The
inspectors evaluated the operators’ ability to implement abnormal and emergency
operating procedures during plant transients with the existing equipment deficiencies.
The review included an evaluation of the cumulative and synergistic effects of the
identified operator workarounds. This inspection activity represented two individual
samples and one cumulative effects of operator workarounds. The following documents
were included in the review:
Procedures and Documents
C
C
C
C
C
b.
Loss of the safety parameter display system, (CR 598870)
Loss of the SPING terminal in the control room and TSC, (CR 603800)
Control Room Deficiency list dated 8/1/04
Operator Workaround list dated 8/30/04
Operator Challenge list dated 9/13/04
Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications (71111.17 - 1 Sample)
a.
Inspection Scope
The inspectors reviewed the system design package and the associated design and
licensing documents for the Unit 1 Fire Protection Cross Tie to Condensate Transfer
System Modification. All functions and design attributes of the modification that could
affect the plant specific SDP worksheets were reviewed. Field implementation activities
were observed and compared to the design requirements and installation standards.
The inspectors reviewed the results of post modification testing. The inspectors also
reviewed the affected procedures and design basis documents to verify that the affected
documents were appropriately updated. Condensate transfer system and fire protection
system were reviewed to verify the modification did not interfere or negatively impact
Enclosure
13
system functions. This inspection activity represented one sample. The following
documents were included as part of the review:
Procedures and Documents
C
C
C
C
C
C
C
b.
Design Change Package (DCP) and Modification Safety Assessment No.
556794, and 556789
NDAP-QA-1220, Modification Process
EO-000-102, “RPV control”
EO-000-113, “Level Power Control”
ON-037-001, “Loss of Condensate Transfer System”
EO-000-114, “RPV Flooding”
ES-013-001, “Fire Protection System Cross-tie to RHR”
Findings
No findings of significance were identified.
1R19 Post Maintenance Testing (71111.19 - 8 Samples)
1.
a.
Routine Review of Post Maintenance Testing
Inspection Scope
The inspectors observed portions of post maintenance testing activities in the field to
determine whether the tests were performed in accordance with the approved
procedures. The inspectors assessed the test’s adequacy by comparing the test
methodology to the scope of maintenance work performed. In addition, the inspectors
evaluated the test acceptance criteria to verify whether the test demonstrated that the
tested components satisfied the applicable design and licensing bases and the
Technical Specification requirements. The inspectors reviewed the recorded test data
to determine whether the acceptance criteria were satisfied. This inspection activity
represented seven samples. The post maintenance testing activities reviewed included:
C
C
C
C
C
C
C
b.
Unit 1 control rod stroke timing after rod 18-23 triple notch withdraw, TP-055-010
Fire diesel replacement of fuel supply solenoids, CR 592724
“E” EDG overhaul, TP-024-149 & MT-024-024
Unit 2 division 2 RHR room cooler and “D” RHR motor maintenance, SO-249B02
Fire system to Unit 2 condensate transfer cross-tie
Control room emergency air supply fan logic test after Agastat relay replacement
WO 522184
Unit 2 SLC flow indicator (FI) FI-24814 replacement AR 599727
Findings
No findings of significance were identified.
Enclosure
14
2.
a.
Unit 2 13.8 KV stepdown transformer 2X250, 2X270 power lead replacement
Inspection Scope
On April 28, 2004, a plant transient resulted from an electrical fault on the 13.8 KV
power cables to transformer 2X270. This fault damaged the 2X270 transformer
enclosure and resulted in an Unusual Event classification. The fault caused a short
duration voltage reduction on the 13.8 KV system and de-energized two 480 volt load
centers. This resulted in the isolation of the condenser offgas system and reduced
condenser vacuum, which required the operators to reduce reactor power. PPL
determined that the fault was the result of insulation breakdown on the 13.8 KV power
cables to transformer 2X270. PPL performed maintenance testing to determine the
health of the transformer. Inspectors observed portions of post maintenance testing
activities in the field to determine whether the tests were performed in accordance with
the approved work documents. The inspectors reviewed the completed work package
and compared it to the work plan and procedures. The inspectors reviewed the
recorded doble test data to determine whether the acceptance criteria were satisfied
prior to placing the transformer back in service. This inspection activity represented one
sample.
b.
Findings
Introduction. A Green NRC-identified finding was identified because PPL did not
complete an evaluation of transformer 2X270 test data as required by work package
instructions and procedure MT-IT-001, “AC Insulation Dielectric Loss and Power Factor
Check.” The inspectors concluded that not completing the evaluation allowed a
degraded transformer (2X270) to be returned to service which resulted in a plant
transient.
Description. The NRC identified that, PPL did not follow maintenance work instructions
and station procedure MT-IT-001, “AC Insulation Dielectric Loss and Power Factor
Check,” which required additional testing and the evaluation of the Unit 2 13.8 KV
stepdown transformer 2X270 doble test data. Specifically, the work instructions and the
initial engineering review determined that testing after the application of heat was
required. Not completing the additional testing and, therefore, not completing the
evaluation of the degraded transformer before energizing the transformer resulted in an
electrical over current fault on the transformer’s primary windings on July 29, 2004. This
fault resulted in the loss of power to various reactor building ventilation fans and
dampers, chillers, a spent fuel pool cooling pump, and the trip and automatic start of
main generator auxiliary system pumps. This fault also resulted in isolation of the
condenser offgas system which reduced condenser vacuum and required the operators
to reduce reactor power to 78 percent.
Following the April 2004 event, PPL initiated work requests to test load center
transformers 2X250 and 2X270 to evaluate the operational readiness of these
transformers prior to reenergizing the transformers. The test results showed a tripling of
the percent power factor over the baseline data, which was an unacceptable change
Enclosure
15
that required further evaluation. The work package test results recommended applying
concentrated local heat to the transformer windings and to retest the transformers to
determine if the unexpected results were due to moisture or other degradation in the
transformer.
PPL applied concentrated heat to dry out any potential moisture in the transformer
windings. However, PPL did not perform a retest or an engineering evaluation to
assess the condition of the transformer, which was required by procedure MT-IT-001,
“AC Insulation Dielectric Loss and Power Factor Check.” Not performing these actions
resulted in energizing a degraded transformer. On July 29, 2004, several hours after
the transformer was energized the primary windings on the transformer faulted which
resulted in a Unit 2 transient.
Analysis. The finding is a performance deficiency because PPL did not follow station
procedure MT-IT-001, “AC Insulation Dielectric Loss and Power Factor Check,” in that
they did no perform an engineering evaluation of an unacceptable change in percent
power factor. Not completing this assessment resulted in a plant transient after this
degraded transformer was energized. This finding is greater than minor because it
adversely impacted the equipment performance attribute of the initiating events
cornerstone and adversely affected the cornerstone objective in that the finding was
associated with an event that upset plant stability. Traditional enforcement does not
apply because the issue did not have any actual safety consequence, or potential for
impacting the NRC’s regulatory function, and is not the result of any willful violation of
NRC requirements. This finding was considered to have very low safety significance
(Green), using Phase 1 of the significance determination process. The issue did not
result in an increase in the likelihood of a loss of coolant accident (LOCA) initiator; and
did not increase the likelihood of a reactor trip and the likelihood that mitigation functions
would be lost. In addition, the finding did not increase the likelihood of a fire or flood
events.
A contributing cause of this finding is related to the Human Performance cross-cutting
area because PPL did not complete the required retest and engineering evaluation of
transformer 2X270 prior to energizing the transformer.
Enforcement. There were no violations of NRC requirements. The inspectors
determined that the finding did not represent a noncompliance because procedure MTIT-001, “AC Insulation Dielectric Loss and Power Factor Check,” is not a procedure that
is referenced among the required procedures listed in NRC regulatory Guide 1.33,
Revision 2, February 1978, Appendix A. Additionally, the 13.8 KV electrical system and
associated balance of plant 480 volt electric distribution system are not safety related.
The related inspection issue was entered into the Susquehanna corrective action
program as CR # 596092. FIN 05000388/2004004-03, “PPL Did Not Retest and
Evaluate Transformer 2X270.”
Enclosure
16
1R22 Surveillance Testing (71111.22 - 8 Samples)
a.
Inspection Scope
The inspectors observed portions of selected surveillance test activities in the control
room and in the field and reviewed the test data results. The inspectors compared the
test result to the established acceptance criteria and the applicable Technical
Specification or Technical Requirements Manual operability and surveillance
requirements to evaluate whether the systems were capable of performing their
intended safety functions. This inspection activity represented eight samples. The
observed or reviewed surveillance tests included:
C
C
C
C
C
C
C
C
b.
Unit 1 Control Rod Coupling Full Indicator Check, SO-156-007
“E” Diesel Generator Monthly Operation Test, SO-024-014
Unit 2 RCIC flow surveillance, SO-250-002
Pump Curve Development for Division II ESW pumps TP-054-066
“C” EDG Start Time Testing, OP-024-005 and OP-024-001
Unit 1 HPCI Flow Surveillance, SO-152-006
ESW Flow Balance, TP-054-076
Unit 2 Shift and Daily Surveillance, SO-200-006
Findings
No findings of significance were identified.
1R23 Temporary Plant Modification (71111.23 - 2 Samples)
a.
Inspection Scope
The inspectors reviewed temporary plant modifications to determine whether the
temporary changes adversely affected system or support system availability, or
adversely affected a function important to plant safety. The inspectors reviewed the
associated system design bases, including the Final Safety Analysis Report (FSAR),
Technical Specifications, and assessed the adequacy of the safety determination
screenings and evaluations. The inspectors also assessed configuration control of the
temporary changes by reviewing selected drawings and procedures to verify whether
appropriate updates had been made. The inspectors compared the actual installations
to the temporary modification documents to determine whether the implemented
changes were consistent with the approved documents. The inspectors reviewed
selected post installation test results to verify whether the actual impact of the temporary
changes had been adequately demonstrated by the test. This inspection activity
represented two samples. The following temporary modifications and documents were
included in the review:
C
Temporary Modification (TMOD) # 574659, Feedwater Heater 3A & 4A Level
Control
Enclosure
17
C
b.
TMOD # 581717, Reactor Building Chiller Discharge Gas Temperature Trip
Elimination
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level (EAL) and Emergency Plan Changes (IP 7111404)
a.
Inspection Scope
A regional in-office review was conducted of licensee-submitted revisions to the
emergency plan, implementing procedures and EALs which were received by the NRC
during the period of April - September 2004. A thorough review was conducted of plan
aspects related to the risk significant planning standards (RSPS), such as
classifications, notifications and protective action recommendations. A cursory review
was conducted for non-RSPS portions. These changes were reviewed against 10 CFR
50.47(b) and the requirements of Appendix E and they are subject to future inspections
to ensure that the combination of these changes continue to meet NRC regulations.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 4, and the applicable requirements in 10 CFR 50.54(q) were used as
reference criteria.
b.
Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06 - 1 Sample)
a.
Inspection Scope
On August 3, the inspectors observed a control room simulator based training event.
The inspectors assessed licenced operator response to simulated plant events and their
use of emergency plan procedures. The inspectors observed PPL’s critique of the
training event to evaluate PPL’s identification of weaknesses and deficiencies
associated with event classification and notifications. The inspectors compared PPL’s
identified findings against the inspectors’ observations to determine whether PPL
adequately identified performance issues. The inspection activity represented one
sample. The inspectors’ review included the following documents and procedures:
C
C
Susquehanna Emergency Plan, revision 45
EP-PS-100, “Emergency Director Control Room”
Enclosure
18
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 9 Samples)
a.
Inspection Scope
The inspector reviewed and assessed the adequacy of the licensee’s internal dose
assessment for any actual internal exposure greater than 50 mrem committed effective
dose equivalent (CEDE).
The inspector examined the licensee’s physical and programmatic controls for highly
activated or contaminated materials (non-fuel) stored within spent fuel and other storage
pools.
The inspector reviewed the licensee’s self-assessments, audits, licensee event reports,
and special reports related to the access control program since the last inspection. The
inspector determined that identified problems were entered into the corrective action
program for resolution.
For repetitive deficiencies or significant individual deficiencies in problem identification
and resolution identified above, the inspector determined that the licensee’s selfassessment activities were also identifying and addressing these deficiencies.
The inspector reviewed licensee documentation packages for all performance indicator
(PI) events occurring since the last inspection.
The inspector selected jobs being performed in radiation areas, airborne radioactivity
areas, or high radiation areas (<1 R/hr) for observation. The inspector reviewed all
radiological job requirements and observed job performance with respect to these
requirements. The inspector determined that radiological conditions in the work area
were adequately communicated to workers through briefings and postings. The jobs
reviewed and observed included the removal and replacement of the filter elements in
the 2B condensate filtration system (CFS) filter.
The inspector discussed with first-line health physics (HP) supervisors the controls in
place for special areas that have the potential to become very high radiation areas
(VHRA) during certain plant operations. The inspector determined that these plant
operations required communication beforehand with the HP group, so as to allow
corresponding timely actions to properly post and control the radiation hazards.
Enclosure
19
This inspection activity represented nine samples. The documents reviewed are
provided in the attachment - “List of Documents Reviewed.”
b.
Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02 - 2 Samples)
a.
Inspection Scope
The inspector reviewed the licensee’s self-assessments, audits, and special reports
related to the As Low As Is Reasonably Achievable (ALARA) program since the last
inspection. The inspector determined that the licensee’s overall audit program scope
and frequency (for all applicable areas under the Occupational Cornerstone) meet the
requirements of 10 CFR 20.1101(c)).
The inspector determined that identified problems are entered into the corrective action
program for resolution. The inspector reviewed dose significant post-job (work activity)
reviews and post-outage ALARA report critiques of exposure performance, and
determined that identified problems are properly characterized, prioritized, and resolved
in an expeditious manner. This inspection activity represented two samples. The
documents reviewed are provided in the attachment - “List of Documents Reviewed.”
b.
Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation (7112103 - 2 Samples)
a.
Inspection Scope
The inspector reviewed licensee self-assessments, audits, and Licensee Event Reports
and focused on radiological incidents that involved personnel contamination monitor
alarms due to personnel internal exposures.
For repetitive deficiencies or significant individual deficiencies in problem identification
and resolution identified above, the inspector determined that the licensee’s selfassessment activities are also identifying and addressing these deficiencies.
The inspector reviewed documents related to the licensee’s processing of
thermoluminescent dosimeters (TLDs) to measure personnel doses of record.
Documents reviewed included the most recent laboratory testing (Personnel Dosimetry
Performance Testing Report dated 9 January 2004) and laboratory audit (On-Site
Assessment 100554-0, February 2003) of the licensee’s program and facility by the
National Voluntary Laboratory Accreditation Program (NVLAP). This inspection activity
Enclosure
20
represented two samples. The documents reviewed are provided in the attachment “List of Documents Reviewed.”
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (71152 - 4 samples)
1.
a.
Identification and Resolution of Problems - Occupational Radiation Safety
Inspection Scope
The inspector selected issues identified in the condition report system for detailed
review. The issues were associated with occupational radiation safety performance
during 2004. The inspector met with the plant radiation protection manager to discuss
these reports. The documented reports for the issues were reviewed to ensure that the
full extent of the issues was identified, an appropriate evaluation was performed, and
appropriate corrective actions were specified and prioritized. This inspection activity
represented one sample.
b.
Findings and Observations
No findings of significance were identified.
2.
a.
Annual Sample Review - Emergency Diesel Generator Maintenance Effectiveness
Inspection Scope
The inspectors reviewed PPL’s initial evaluation and associated corrective actions for 50
condition reports (CRs) related to emergency diesel generator (EDG) problems between
November 1999 and June 2004. This sample was selected due to the potential for an
adverse trend with regard to resolving potentially repetitive EDG maintenance issues.
The review evaluated PPL’s threshold for identifying and resolving problems. This
inspection activity represented one sample. The documents reviewed during this
inspection are listed in the Supplemental Information Attachment (back of this report).
b.
Findings and Observations
No findings of significance were identified.
Observations
The inspectors evaluated PPL's problem identification and resolution efforts for EDG
issues for the previous five years. The inspectors determined that, in general, PPL was
Enclosure
21
effective at identifying problems and placing them in the corrective action program. PPL
evaluated many of the issues as only affecting the "test mode" of EDG operations, and
concluded that the EDG would have been able to perform its safety function. As a
result, PPL was slow to identify adverse trends in EDG maintenance effectiveness; PPL
identified an adverse trend in EDG reliability in July 2003. The inspectors noted
examples where problem resolution was narrowly focused or lacked sufficient technical
analysis to prevent recurrence. As a result, PPL's EDG maintenance effectiveness
remained weak during the 5-year period reviewed.
The inspectors identified a high failure rate for EDG post maintenance tests (PMTs)
performed following EDG overhauls (9 failures out of 13). In each case, additional
unplanned EDG unavailability and TS LCO time was incurred. In one instance, the 72hour TS LCO expired, and a reactor shutdown was commenced.
In addition to a high post-overhaul test failure rate, the inspectors identified multiple
examples, during the 5-year period, where weak maintenance practices resulted in
EDGs becoming inoperable or unavailable during surveillance testing or standby
service. The most significant issues included:
C
C
C
C
C
C
C
January 2004, with the "B" EDG in standby, an emergency service water leak
identified that half the bolts on the lube oil cooler heat exchanger were loose;
attributed to a degraded gasket. Subsequent investigation identified multiple
fastener issues on all five EDGs, and resulted in an NRC Special Inspection (IR
50-387,388/2004-007).
January 2004, the "A" EDG failed a surveillance test due to an oil spray, which
identified loose governor mounting bolts; attributed to improper torque when last
replaced during the August 2003 overhaul.
July 2003, with the "C" and "D" EDGs in standby, PPL identified that the
generator outboard bearings had been improperly torqued when last assembled;
attributed to an inadequate procedure, which was not previously identified
because the mechanics relied on skill-of-the-craft for bearing assembly, not
procedural step-by-step compliance.
May 2003, the "A" EDG tripped during a surveillance test; attributed to a loose
wire on a relay base. The relay base had been replaced in October 2002, as a
corrective action for a similar EDG trip.
May 2003, the "C" EDG failed a surveillance test because load could not be
increased above 100 kW. This was attributed to a loose connection in the test
circuitry.
March 2003, the "D" EDG failed a surveillance test when the mechanical
governor became disconnected from the fuel rack linkage because a connecting
bolt fell out; initially, PPL attributed the failure to excessive vibration. The
inspectors subsequently identified that the bolt had not been torqued when the
governor was last replaced during the June 2002 overhaul.
January 2003, the "B" EDG tripped during a PMT because a jumper had not
been reinstalled on a motor-operated potentiometer, which had just been
replaced.
Enclosure
22
PPL has recently downgraded the EDG overall system health because of a negative
trend in system performance. PPL developed a plan to improve the overall system
health and documented this in CR 580429. This plan “The Road to Green,” will
addresses maintenance issues by creating a multi-discipline Diesel Review Team which
will address issues related to the maintenance of EDG components.
3.
a.
Annual Sample Review - Diesel Driven Fire Pump Lack of Engine Cooling
Inspection Scope
The inspector reviewed PPL’s evaluations and associated corrective actions for selected
condition reports related to the diesel fire pump (DFP). The DFP was selected due to
the numerous condition reports written on the component and the risk insights with
respect to late reactor vessel injection. The inspectors completed a detailed walkdown
of the DFP engine and controls. The inspectors reviewed the FSAR, Operating
procedures, Fire Protection Review Report, and the Individual Plant Evaluation (IPE) to
ensure the equipment is operated and maintained consistent with the design and risk
significance of the system. Inspectors also reviewed equipment vendor manual
instructions to assess the significance of operation beyond established limits. This
inspection activity represented one sample. The documents reviewed are provided in
the report attachment.
b.
Findings and Observations
Introduction. The NRC identified a Green finding for failing to adequately evaluate and
correct high engine operating temperatures and repetitive overheating of the DFP after
shutdown, in accordance with NDAP-QA-0702, “Action Request and Condition Report
Process.” For over three months, PPL did not evaluate or initiate specific action to
correct the high temperature condition which resulted in repetitive maintenance and
unavailability time for the DFP.
Description. During May, June, and July 2004, the inspectors observed several
instances where the DFP was operated with cooling water temperatures exceeding the
vendor’s recommended high temperature of 195EF. The inspectors also observed
instances in which the vendor’s maximum engine coolant temperature of 200EF was
exceeded after the engine was shutdown. The observations are listed below.
C
C
C
C
May 2004, during post maintenance testing the inspectors observed the DFP
engine overheat to greater than 220EF and the overflow of coolant immediately
following engine shutdown.
June 1, 2004, the DFP engine cooling water temperatures were above the
operating procedure limit of 185EF.
July 14, 2004, during troubleshooting and post maintenance testing, for a failure
of the DFP to automatically start, inspectors observed high engine coolant
temperatures greater than 195EF.
July 21, 2004, the inspectors again observed high engine operating
temperatures of 199EF and DFP engine temperature increased to 225EF when
Enclosure
23
shutdown with the overflow of approximately two pints of engine coolant onto the
engine. PPL did not initiate corrective action reports for the July 21, 2004 event,
the plant operators referenced previously issued CRs in the work package.
Prior to the NRC inspectors questioning DFP availability and reliability, PPL did not
evaluate the cause of the elevated engine temperatures and had not taken effective
actions to correct the condition or evaluate the severity of the engine coolant overflow
and engine overheating following shutdown. The DFP was removed from service for
corrective maintenance numerous times without any action to restore engine cooling
capability. The vendor manual states that if the engine is stopped suddenly, the
turbocharger temperature can rise as much as 100EF. The Turbocharger is cooled by
lubricating oil which is cooled by the engine coolant (water/glycol). A high degree of
heat in the turbocharger can cause seizer of bearings, burned O-ring oil seals and the
distortion of the bearing housing. Inspectors also determined that the overflow of
engine coolant over the engine components including the starting solenoids has the
potential to reduce component reliability.
On July 21, 2004, the inspectors discussed, with PPL, the repeated coolant overflow
and engine overheating following shutdown as well as the elevated engine temperatures
above procedure limits during pump operation. On July 27, 2004, PPL took corrective
action to change the pressure control valve setpoint to increase cooling water flow to the
DFP engine and bring engine temperatures back into the normal operating band. The
cause of the inadequate cooling water flows was attributed to previously establishing a
pressure control valve setpoint that was based on the temperature conditions in
November but was not adequate for summer season operation. There was also
instrument drift found in the pressure control valve setpoint that decreased flow to the
engine heat exchanger.
Analysis. The finding is a performance deficiency because PPL failed to correct a
degraded equipment condition that was captured in the corrective action system,
specifically, the elevated temperatures and engine overheating of the DFP. Traditional
enforcement does not apply because the issue did not have any actual safety
consequences or potential for impacting the NRC’s regulatory function and was not the
result of any willful violation of NRC requirements or PPL procedures. This issue is
greater than minor because the DFP is utilized as a mitigation system and elevated
temperatures, coolant overheating and coolant discharge, reduced the reliability and the
availability of the system. Thus, the finding affected the Mitigating Systems cornerstone
objective.
This finding was assessed in accordance with NRC Manual Chapter 0609, Appendix A,
Attachment 1, "Significance Determination Process (SDP) for Reactor Inspection
Findings for At-Power Situations," and was determined to be of very low safety
significance (Green) based on a Phase 1 analysis. The inspectors determined that this
condition did not represent an actual loss of the equipment designated as risk-significant
for greater than 24 hours, because there was no loss of the reactor vessel late injection
capability. Additionally, the finding does not increase the potential or risk of a seismic
event, flood or severe weather event.
Enclosure
24
A contributing cause of this finding is related to the Problem Identification and
Resolution (PI&R) cross-cutting area. PPL did not sufficiently evaluate the condition to
identify and correct the reduced cooling water flow to the DFP engine. This resulted in
ineffective corrective actions because the DFP was removed from service several times
without taking action to correct the DFP high engine coolant temperature issue.
Enforcement. No NRC requirements were violated. The Diesel Driven Fire pump is not
safety related and thus not covered by 10 CFR 50, Appendix B. PPL failed to correct a
degraded condition on a system utilized to mitigate station risk. The issues associated
with this finding are entered into the corrective action program under CR 594877.
FIN 05000387, 388/2004004-04, “Diesel Driven Fire Pump Lack of Engine Cooling.”
4.
a.
Annual Sample Review - N2J and N1B Nozzle Weld Indications
Inspection Scope
The inspectors reviewed conditions reports associated with a crack at the weld root in
each of the N1B and N2J safe end to nozzle welds in the Unit 1 recirculation system.
These cracks were located during a scheduled Inservice Inspection (ISI) using
computer based ultrasonic examination techniques. The condition reports were
reviewed to ensure a complete and accurate identification of the issue, an appropriate
root cause evaluation was performed, extent of condition was considered, and corrective
actions were implemented. In addition to the condition reports, the inspector reviewed
design calculations, welding procedures, and process qualifications used to support the
repair effort.
The inspector also conducted interviews with personnel responsible for the
nondestructive processes used to identify the indications, performance of the root cause
evaluation, development of the repair plan, implementation of the plan and verification
the repair met the specified acceptance criteria. This inspection activity represented
one sample. The documents reviewed during this inspection are listed in the
Supplemental Information Attachment (back of this report).
b.
Findings and Observations
No findings of significance were identified.
Observations
PP&L discovered significant indications in the Unit 1 N1B and N2J safe end to reactor
vessel nozzle welds through ultrasonic testing performed during refueling outage 13RIO
(Spring 2004). Ultrasonic test data was evaluated and the indications identified were
characterized as cracks which had initiated at the root of the welds and propagated into
the weld thickness. Both indications were found to be acceptable within the ASME
Section XI, IWB 3640 design requirements. However, PP&L elected to install a weld
overlay on both welds to alleviate any concerns regarding crack growth with potential for
penetration through the pressure boundary during subsequent operating cycles.
Enclosure
25
The inspector noted that these welds had been previously inspected at established
ASME Code intervals using both surface (penetrant test) and volumetric (ultrasonic test)
techniques. Tests were performed using qualified techniques and personnel as
required by the ASME Section XI Code. Although indications were identified during
these previous inspection periods, a number of the indications were not identified, sized
or characterized as “cracks.” As a result of the failure to identify and characterize these
indications as linear or, as cracks, no further flaw evaluations were performed since the
indications were reported as “non-relevant” or “root geometry.” This was consistent with
the limitations of the equipment and examination techniques in effect at the time of
these previous examinations.
Ultrasonic examinations performed beginning in 13RIO (April 2004) were completed
using Performance Demonstration Initiative (PDI) qualified procedures and examiners.
The procedures, equipment and examiners qualified for this testing represent the most
current available technology and equipment coupled with rigorous training and testing of
personnel to assure optimum results. This examination system was used for the first
time during 13RIO to examine the N1B and the N2J welds and had not been available
for previous examinations.
The inspector reviewed the previous results of surface and volumetric tests performed
on both the N1B and the N2J welds in an effort to determine why the indications were
not identified, characterized and evaluated as “cracks” prior to the Spring 2004
examination. Based on this review, the inspector concluded that the failure to identify
and adequately characterize the indications in the N1B and N2J nozzle to safe-end
welds during previous examinations was the result of a combination of procedure,
equipment and examiner limitations. However, the inspector noted that at the time the
previous examinations were performed, the procedures, equipment and examiners were
qualified to the requirements of ASME Section XI. Examinations were implemented and
results interpreted in accordance with the Code requirements in effect at the time of the
examinations. Indications noted were found to meet the acceptance criteria provided in
the ASME Section XI Code.
5.
a.
Routine PI&R Review
Inspection Scope
The inspectors reviewed selected condition reports (CRs), as part of the routine
baseline inspection documented in this report. The CRs were assessed to verify
whether the full extent of the various issues were adequately identified, appropriate
evaluations were performed, and reasonable corrective actions were identified. The
inspectors evaluated the CRs against the requirements of NDAP-QA-0702, "Action
Request and Condition Report Process," and 10 CFR 50, Appendix B. During this
inspection period, the inspectors performed a screening review of each item that PPL
entered into their corrective action program, to assess whether there were any
unidentified repetitive equipment failures or human performance issues that might
warrant additional follow-up.
Enclosure
26
b.
Findings and Observations
No findings of significance were identified.
6.
Cross-References to PI&R Findings Documented Elsewhere
Section 1R12 describes a finding where PPL had numerous condition reports that
discussed problems with the equipment floor drain system (EFDS) but did not identify
the design significance of this system and did not specify adequate corrective actions to
maintain the functionality of the system.
Section 4OA2.3 describes a finding where PPL did not sufficiently evaluate the
overheating of the diesel fire pump (DFP) to identify and correct this degraded condition.
4OA3 Event Follow-up (71153 - 3 Samples)
1.
(Closed) LER 05000387/2004001-00 Automatic Actuation of ‘A’ Emergency Diesel
Generator when Operator Removed Incorrect 4.16 KV Bus Fuse
On March 7, 2004, the ‘A’ Emergency Diesel Generator (EDG) started due to a detected
under-voltage condition on the 1A201 4.16 KV Engineering Safeguards bus. The bus
undervoltage condition was created when an in-plant operator removed fuses from an
incorrect bus breaker cubicle during equipment alignment.
The inspectors reviewed the event and documented their assessment in NRC Inspection
Report 50-387,388/2004-002, Section 1R14 Item 2. The documented assessment
included a very low safety significance (Green) NCV that was identified because a nonlicensed plant operator did not implement the electrical system shutdown procedure as
written. (NCV 05000387,388/2004002-01).
The inspectors reviewed the related information and no additional findings were noted.
This issue was documented in PPL's corrective action program as CR 555676. This
LER is closed.
2.
(Closed) LER 05000387/2004002-00 Loss of Safety Function - Control Structure
Chillers Inoperable due to Blank Flanges Being Installed in Wrong Emergency Service
Water Loop (Common)
On March 9, 2004, PPL identified that both control room emergency outside air supply
system (CREOAS) were inoperable because the “A” and “B” loops of emergency service
water (ESW) system that supplied the cooling to the CREOAS system were inoperable
and the associated TS limiting condition of operation had not been entered. Specifically,
on March 5, 2004, during a refueling outage with irradiated fuel movements in progress,
the emergency supply of cooling to “A” CREOAS was out of service when maintenance
personnel incorrectly removed a blank flange in the Unit 1 “B” ESW loop. Removal of
the blank flange resulted in the “B” ESW piping being in an unanalyzed seismic
condition and was declared inoperable in accordance with TS. During the one hour
Enclosure
27
period, both CREOAS trains were inoperable because the “A” ESW supply was not
available and the “B” ESW supply was not fully seismically qualified. Fuel moves were
not suspended as required by TS 3.7.3.F and 3.7.4.E. PPL determined the cause to be
a lack of a pre-job brief and a poor turnover between maintenance shifts. Corrective
actions included returning the ESW and CREOAS systems to an operable status and
including this event into station training.
This finding is more than minor because it affected the human performance attribute of
the containment barrier cornerstone. The finding was considered to have very low
safety significance (Green) using SDP Phase 1 screening. Because this finding only
represented a degradation of the radiological barrier function provided for the control
room because the CREOAS system did not have a fully seismically qualified safetyrelated supply of cooling water for one hour. This licensee-identified finding involved a
violation of TS 3.7.3 and 3.7.4. The enforcement aspects of this violation are discussed
in Section 4OA7. This LER is closed.
3.
(Closed) LER 05000387/2004003-00 Manual Scram Following Turbine High Vibration
On April 21, 2004, Unit 1 was operating at 17% power when the reactor was manually
scrammed and condenser vacuum was broken to rapidly reduce turbine speed in
response to main turbine bearing high vibrations. The high vibration was the result of
component rubs that were expected, in part, due to the newly installed turbine and not
setting the turbine vibration monitoring limits and turbine trip limit low enough to prevent
high vibrations. PPL revised OP-1(2)93-001, “Main Turbine Operation,” to include clear
guidance on turbine vibration monitoring and lower vibration limits to allow quicker
detection of elevated turbine vibration. The LER was reviewed by the inspectors and no
finding of significance were identified. PPL documented this condition in CR 573728.
This LER is closed.
4OA4 Cross Cutting Aspects of Findings
Cross-References to Human Performance Findings Documented Elsewhere
Section 1R19 describes a finding where PPL did not complete a required retest and
engineering evaluation of a degraded transformer prior to returning the transformer to
service.
4OA5 Other
1.
a.
Operation of an Independent Spent Fuel Storage Installation (60855)
Inspection Scope
The inspector observed selected spent fuel dry cask loading operations for dry shielded
canister (DSC) 29 which were conducted in accordance with procedure ME-0RF-023,
“Dry Fuel Storage - 61BT Dry Shielded Canister.” Specifically, installation and welding
of the inner top cover, canister blowdown, and initial vacuum drying operations were
Enclosure
28
observed. Vacuum drying and Helium leak test results for DSCs 26, 27, and 28 were
reviewed with respect to nuclear horizontal modular storage (NUHOMS) 52B and 61BT
Technical Specifications (TS) 1.2.2, 1.2.4 and 1.2.4a criteria, respectively. Radiological
work practices and exposure rates were discussed with technicians responsible for
ongoing work. Conformance to the requirements of TS 1.2.11 and 1.2.12, “Transfer
Cask Dose Rates” and “Maximum DSC Removable Surface Contamination,” criteria,
was evaluated. Personnel exposures were reviewed and radiation work permit (RWP)
2004-0200, “Dry Fuel Storage Activities On The Refuel Floor” was examined.
The inspector discussed with cognizant licensee representatives the procedural controls
in place that ensured only designated fuel assemblies were properly selected and
loaded into NUHOMS 52B and 61BT Casks. A review of the spent fuel assembly move
sheets and verification records required by RE-081-042, “Fuel and Core Component
Transfer Authorization Sheet (FACCTAS) Preparation Guidelines,” and RE-081-043,
“Selection and Monitoring of Fuel for Dry Storage,” was conducted. The inspector
observed a video tape of final fuel configuration in NUHOMS 52B DSCs 26 and 27, and
NUHOMS 61BT DSCs 28 and 29, which indicated fuel assembly serial numbers. Fuel
characteristics including enrichments, burn-up, post irradiation cooling time, heat
generation, and known structural defects, were reviewed and evaluated against the
NUHOMS TS 1.2.1 limits.
The inspector reviewed 10CFR72.48 safety evaluations generated since the last spent
fuel transfer campaign in 2002, including a 72.48 screen ( 7248-01-117, “Convert to
61BT Dry Shielded Canisters”) which evaluated the effect of Amendment 4 to Certificate
of Compliance 1004 on the NUHOMS Storage System at Susquehanna. The inspector
also reviewed Evaluation SE 72-1778: Use of Wrong Gas in DSC Canister.
Training and qualifications of selected personnel involved with dry cask storage work
was reviewed to ensure adherence to the general license criteria in 10CFR72.212(b)(6).
This review included personnel responsible for rigging and cask handling, welding,
vacuum drying, helium backfill operations, and helium leak testing. Inclusion of
operating experience was also evaluated.
Effectiveness of licensee management oversight, quality assurance (QA) audits and self
assessments, and corrective action program as applied to the dry cask storage program
was evaluated. A management readiness review, conducted prior to the start of this
campaign and approved by the station PORC, was reviewed. The inspector discussed
operational oversight with a QA inspector on the refuel floor during operational activities
associated with DSC 29.
A tour of the ISFSI pad and enclosed area was conducted to ensure proper
housekeeping and conformance to combustible loading limits.
b.
Findings
No findings of significance were identified.
Enclosure
29
2.
(Closed) SL-III Violation 05-387, 388/2002005-02, Spent Fuel Cannister Filled with
Wrong Gas
On July 26, 2002, PPL filled a spent fuel storage canister (DSC) with Argon and Helium
gases instead of using all Helium gas as required by the NUHOMS Technical
Specification. A Severity Level III violation (with no civil penalty) was issued to PPL for
this event in inspection report 2002-005. During the inspection, PPL’s corrective actions
to prevent the use of the Argon gas were reviewed. The inspector verified procedure
changes were completed and training on the revised procedures conducted prior to the
start of the next dry cask storage campaign. The inspector observed equipment set up
on the refuel floor and determined that adequate controls were instituted to prevent
recurrence of the violation. This item is considered closed.
4OA6 Meetings, Including Exit
On October 7, 2004, the resident inspectors presented the inspection results to Mr. B.
McKinney, Vice President - Nuclear Operations, and other members of your staff, who
acknowledged the findings. The inspectors confirmed that proprietary information was
not provided or examined during the inspection.
4OA7 Licensee-identified Violations
The following violation of very low safety significance (Green) was identified by PPL and
is a violation of NRC requirements which meet the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.
C
TS 3.7.3.F and 3.7.4.E requires that fuel moves be immediately terminated when
both trains of CREOAS are inoperable. Contrary to this on March 5, 2004, both
trains of CREOAS were determined to be inoperable due to a maintenance error
on a support system and fuel moves were not suspended. This was identified in
PPL’s corrective action program as CR 556923. This finding is of very low safety
significance because even though the “B” ESW system was not fully seismically
qualified, PPL determined that the system remained capable of supplying water
to the necessary heat loads during this period.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
A-1
SUPPLEMENTAL INFORMATION
KEY POINT OF CONTACT
Licensee Personnel
S. Ingram, Senior Health Physicist
E. McIlvaine, Jr., Health Physics Operations Foreman
V. Schuman, Radiological Operations Supervisor
L. Wolf, Health Physics Operations Foreman
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
050000387/2004004-01
URI
Equipment Hatch Floor Plugs are not Watertight as
Indicated in the FSAR
05000387/2004004-02
NCV
Reactor Building Floor and Equipment Drains Not
Fully Scoped into the Maintenance Rule
05000388/2004004-03
FIN
PPL Did Not Retest and Evaluate Transformer
2X270
05000387, 388/2004004-04
FIN
Diesel Driven Fire Pump Lack of Engine Cooling
05000387/2004001-00
LER
Automatic Actuation of ‘A’ Emergency Diesel
Generator When Operator Removed Incorrect
4.16KV Bus Fuses (Common)
05000387/2004002-00
LER
Loss of Safety Function - Control Structure Chillers
Inoperable due to Blank Flanges Being Installed in
Wrong Emergency Service Water Loop (Common)
05000387/2004003-00
LER
Manual Scram Following Turbine High Vibration
05000387, 388/2002005-02
VIO
Spent Fuel Cannister Filled with Wrong Gas.
Opened and Closed
Closed
Attachment
A-2
LIST OF BASELINE INSPECTIONS PERFORMED
7112101
7112102
7112103
Access Control
ALARA Planning and Controls
Radiation Monitoring Instrumentation
2OS1
2OS2
2OS3
LIST OF DOCUMENTS REVIEWED
(Not Referenced in the Report)
1R06: Flood Protection Measures
Internal Flood Protection
EC-FLOD-0500,
EC-FLOD-0001,
EC-012-6047
EC-RISK-0539
PPL DBD 0010
Section 2.10,
Review of MISC-037 calculation for Determining Correct Flood Depths for
Reactor Building Elevation 683 Foot
Moderate Energy Pipe Breaks - Floods
Removal of Floor Plugs in Conditions 4 or 5
Internal Flooding Analysis for PRA
Internal Flooding Functional Requirements
FSAR Section 3.4.1 Water Level Design / Flood Protection
FSAR Section 3.6
Plant Design for Protection Against Postulated Piping Failures in Fluid
Systems Outside Containment
FSAR Section 3.12 Separation Criteria for Safety-Related Mechanical and Electrical Power
Equipment
FSAR Section 6.3
Emergency Core Cooling Systems
CR 600070
CR 600070
Operability Determination, Rev. 0
Operability Determination, Rev. 1
PPL Evaluation of Water Intrusion into ECCS Rooms
Regulatory Guide 1.102, Rev. 1, Flood Protection for Nuclear Power Plants
2OS1: Access Control to Radiologically Significant Areas
Section 2: Radiation Safety
Audit Reports
AR 542112, Y2003 Trend Evaluation
AR 551986, Radiation Protection Activities Observed during the Unit 1 13th RIO
Attachment
A-3
Self-Assessments
HPS-03-003, Training Program Self-Assessment
HPS-03-04, Comprehensive RP Assessment (10CFR20.1103c)
HPS-03-05, Assessment of Batelle PNL Irradiation Services
HPS-03-06, Radioactive Material Control
Condition Reports
559453, 569363, 569364, 569688, 573291, 574278, 577880, 578045, 578630, 579754,
580133, 580364, 580984, 581740, 581960, 582250, 582256, 582262, 582272, 582277,
582279, 582660, 589439, 590163, 590181, 590286, 591947, 593065
4OA2: Identification and Resolution of Problems
Annual Sample Review - EDG Maintenance Effectiveness
Procedures
NDAP-QA-0702, "Action Request and Condition Report Process"
IC-024-001, "Electronic Tune-up for Diesel Generator Governors"
Technical Specification (TS) 3.8.1, surveillance requirements, and TS Basis
Miscellaneous
"A" EDG Speed Control Troubleshooting Report, MPR Associates Inc., dated December 13,
1999
"Unexpected "A" Diesel Generator Operation Root Cause Analysis Report," ANNA Inc., dated
January 18, 2000
"Memorandum to ESI Source Inspection Report 2000-20," Engine Systems Inc., dated
July 18, 2000
"Laboratory Examination of Failed Bearing Temperature Sensor for Solder Coverage," PPL
Metals Lab Report 2001-094, dated September 18, 2001
"Failure Analysis of Cracked Lower Cylinder Liner Expansion Sleeve," PPL Metals Lab Report
2002-003, dated January 18, 2002
"EDG Governor Component Failure Analysis Report MPR-2513," MPR Associates Inc., dated
March 26, 2003
"DG Reliability Report," PPL Trend Review Team, dated September 11, 2003
Condition Reports
585913, 579510, 564942, 544874, 543502,
512609, 508825, 505495, 488000, 487476,
454959, 446398, 445656, 445506, 445489,
428743, 408025, 376231, 362560, 353773,
273269, 270141, 269851, 265604, 265556,
543496, 543172, 532012, 532461, 512613,
484120, 478691, 475852, 474234, 460227,
445379, 445361, 445171, 445139, 430146,
353093, 340936, 316629, 290686, 274297,
265475, 218433, 216004, 212940, and 212844
Attachment
A-4
Annual Sample Review - N2J and N1B Nozzle Weld Indications
Condition Reports,
561319, 567886, 561319, 567886, 573890, 592892, 58437, 573883, 573882, 564420, 563116,
573889, 573893, 562164, 573887, 562886, 567886, 570394, 565923, 566347
Miscellaneous
Level 1 Root Cause Evaluation for Unit 1N1B and N2J Nozzle Flaws
Overlay Weld Procedure WPS 03-08-T-801 Rev 0
Weld Overlay Repair Plan SIR-04-038 Rev 0
Summary Reports of UT Exam of Unit 1 N1B and N2J Nozzle to Safe End Weld
MSIP Performance and Verification Record for Unit 1 N1B and N2J Welds
Engineering Calculation EC-062-1083 Rev 0
PCN 2004-6463, 6465, N1B and N2J Overlay Thickness Verification
DCP 567820, Rev 0, Unit 1 Weld Overlay Installation Instructions
LIST OF ACRONYMS
ALARA
ASME
CEDE
CFR
CFS
CIG
CR
CREOAS
DFP
DSC
EAL
ECCS
EDG
EFDS
EP
EF
FACCTAS
FI
ESW
FSAR
HP
HPCI
HRA
HSM
ICM
IMC
IPE
As Low As Is Reasonably Achievable
American Society of Mechanical Engineers
Committed Effective Dose Equivalent
Code of Federal Regulations
Condensate Filtration System
Containment Instrument Gas
Condition Report
Control Room Emergency Outside Air Supply
Diesel Fire Pump
Dry Shielded Canister
Emergency Action Level
Emergency Core Cooling Systems
Emergency Diesel Generator
Equipment Floor Drain System
Emergency Preparedness
Degrees Fahrenheit
Fuel and Core Component Transfer Authorization Sheet
Flow Indicator
Emergency Service Water
[SSES] Final Safety Analysis Report
Health Physics
High Pressure Coolant Injection
High Radiation Area
Horizontal Storage Module
Interim Compensatory Measures
[NRC] Inspection Manual Chapter
Individual Plant Evaluation
Attachment
A-5
ISFSI
KV
kW
LCO
LOCA
LER
MSIV
NCV
NDAP
NPO
NRC
NUHOMS
NVLAP
PCO
PI
PI&R
PMT
PPL
PSV
QA
RB
RCIC
RG
RHR
RPV
RSPS
RWCU
RWP
SDP
SLC
SOW
SPDS
SPING
SSC
SSES
TBCCW
TLD
TMOD
TS
URI
VHRA
WO
Independent Spent Fuel Storage Installation
Killovolts
Killowatts
[TS] Limiting Condition for Operation
Loss of Coolant Accident
Licensee Event Report
Main Steam Isolation Valve
Non-cited Violation
Nuclear Department Administrative Procedure
Nuclear Plant Operator
Nuclear Regulatory Commission
Nuclear Horizontal Modular Storage
National Voluntary Laboratory Accreditation Program
Plant Control Operator
[NRC] Performance Indicator
Problem Identification and Resolution
Post Maintenance Test
PPL Susquehanna, LLC
Pressure Safety valve
Quality Assurance
Reactor Building
Reactor Core Isolation Cooling
[NRC] Regulatory Guide
Residual Heat Removal
Reactor Pressure Vessel
Risk Significant Planning Standards
Reactor Water Cleanup
Radiation Work Permit
Significant Determination Process
Standby Liquid Control
System Outage Window
Safety Parameter Display System
System Site Particulate Iodine and Nobel Gas
Structure, System, or Component
Susquehanna Steam Electric Station
Turbine Building Closed Cooling Water
Thermoluminescent Dosimeter
Temporary Modifications
Technical Specification
Unresolved Issue
Very High Radiation Area
Work Order
Attachment